RBG-47157, License Amendment Request, Changes to Technical Specification 3.3.6.1, Primary Containment and Drywell Isolation Instrumentation

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License Amendment Request, Changes to Technical Specification 3.3.6.1, Primary Containment and Drywell Isolation Instrumentation
ML11214A093
Person / Time
Site: River Bend Entergy icon.png
Issue date: 07/27/2011
From: Roberts J
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RBG-47157
Download: ML11214A093 (33)
v • d • e


Text

Entergy Operations, Inc.

River Bend Station r5485 U.S. Highway 61 N n

St. Francisville, LA 70775 Tel 225-381-4374 Jerry C. Roberts Director, Nuclear Safety Assurance RBG-47157 July 27, 2011 U.S, Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555,

SUBJECT:

License Amendment Request Changes to Technical Specification 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation" River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47

Dear Sir or Madam:

In accordance with the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR), Entergy Operations, Inc. (Entergy) is submitting a request for a amendment to the Technical Specifications (TS) for River Bend Station (RBS), Unit 1. A change is proposed to Technical Specification (TS) 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation" to revise the allowable value setpoints for the Main Steam Tunnel Temperature functions 1.e, 3.f and 4.h.

This setpoint revision is based upon a revision to the analytical limit calculation. The change will provide additional margin for elevated temperatures in the Main Steam Tunnel

- North during the summer reliability period.

In addition to the changes to the RBS Technical Specifications the enclosed Emergency Plan changes are also submitted for NRC staff review and approval as required by 10 CFR 50.54(q), 50.4, and 50.90. The proposed change modifies the station's Emergency Action Levels (EAL) in support of the proposed changes to TS 3.3.6.1.

The proposed change to the EALs was evaluated against the criteria of 10 CFR 50.47, 10 CFR 50, Appendix E and other NRC guidance documents. This EAL change has been evaluated in accordance with 10 CFR 50.54(q) and the guidance contained in NRC RIS 2005-02, and the evaluation has deemed NRC prior approval is required. The requested change is acceptable in that there is reasonable assurance that adequate protective measures can and will be taken in the event of a radiological emergency.

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RBG-47157 Page 2 of 3 The proposed changes have been evaluated in accordance 10 CFR 50.92(c) and it has been determined that the changes involve no significant hazards considerations. provides the No Significant Hazards Consideration for the change. provides a description of the proposed change. Attachment 2 provides the existing TS pages marked up to show the proposed change. Attachment 3 provides the existing TS Bases pages marked up to show the proposed change (for information only). provides Justification of Emergency Plan Emergency Action Level change, a mark-up of the latest Emergency Action Levels and a revised copy of Emergency Action Levels. Attachment 5 provides a summary of the regulatory commitments made in this submittal.

This change has been reviewed and approved by the Onsite Safety Review Committee (OSRC).

Although this request is neither exigent nor emergency, your prompt review is requested.

Once approved, the amendment shall be implemented within 60 days. If you have any questions or require additional information, please contact Mr. Joseph A. Clark at (225) 381-4177.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 27, 2011 Sincerely, JCR/JAC/bmb Attachments:

1. Analysis of Proposed Technical Specification Change
2. Proposed Technical Specification Changes (mark-up)
3. Changes to Technical Specification Bases Pages - For Information Only
4. Justification of Emergency Plan Emergency Action Level change
5.

List of Regulatory Commitments cc: Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 7601 1-4125

RBG-47157 Page 3 of 3 NRC Senior Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS 0-8 B1 Washington, DC 20555-0001 Mr. Jeffrey P. Meyers Louisiana Department of Environmental Quality Office of Environmental Compliance Attn:OEC - ERSD P. 0. Box 4312 Baton Rouge, LA 70821-4312 3

Attachment I RBG-47157 Analysis of Proposed Technical Specification Change RBG-47157 Page 1 of 11

1.0 DESCRIPTION

The proposed amendment would revise River Bend Station (RBS), Unit 1 Technical Specification (TS) 3.3.6.1;" Primary Containment and Drywell Isolation Instrumentation,"

allowable value setpoints contained in TS 3.3.6.1-1 items; i.e "Main Steam Tunnel Temperature-High," 3.f "Main Steam Line Tunnel Ambient Temperature-High," 4.h "Main Steam Line Tunnel Ambient Temperature-High."

In addition to the changes to the Technical Specification allowable value setpoints a corresponding change will be made to increase the nominal trip value in the Technical Requirements Manual (TRM) functions identified above.

This request is to change the Technical Specification allowable value and the associated change to the Emergency Action Level (EAL) under 10 CRF 50.90. The changes to the setpoint nominal values and the use of GOTHIC are addressed under 10 CFR 50.59.

2.0 PROPOSED CHANGE

The Technical Specifications (TS) functions which this proposed change addresses the new proposed allowable value of 5 183.0 °F are the following:

1. TS Table 3.3.6.1-1 Function 1.e - Main Steam Line Isolation -Main Steam Tunnel Temperature-High
2. TS Table 3.3.6.1-1 Function 3.f-Reactor Core Isolation Cooling (RCIC) System Isolation -Main Steam Line Tunnel Ambient Temperature-High
3. TS Table 3.3.6.1-1 Function 4.h - Reactor Water Cleanup (RWCU) System Isolation -Main Steam Line Tunnel Ambient Temperature-High contains a markup of the proposed Technical Specification pages.

In addition to the identified changes to the Technical Specifications above, the BASES will be revised upon implementation to include the following information based upon TSTF-493 BASES revision to NUREG-1434, SR 3.3.6.1.5, to include the following statement; For Functions 1.e, 3.f and 4.h there is a plant specific proqram which verifies that the instrument channel functions as required by verifying the as-left and as-found setting are consistent with those established by the setpoint methodology. contains a markup of the proposed Technical Specification BASES page.

3.0 BACKGROUND

During the summer months the Main Steam Line Tunnel area ambient temperatures in conjunction with minor steam leaks in the MSL Tunnel have approached the trip setpoints. These setpoints are designed to detect a 25 gpm leak in the Main Steam RBG-47157 Page 2 of 11 Tunnel - North area and initiates plant isolations. The proposed changes will maintain the design leak detection criteria while providing additional margin for plant operation.

Technical Specification 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation" contains the Allowable Values for isolation instrumentation that cause initiation of primary containment and reactor coolant pressure boundary (RCPB) isolation. Functional diversity is provided by monitoring a wide range of independent parameters. One of the parameters monitored to provide isolation of the main steam (MS) lines, reactor core isolation cooling (RCIC) lines and reactor water clean up (RWCU) lines is Main Steam Tunnel Temperature - High.

In order to provide additional margin for elevated temperatures in the Main Steam Tunnel - North during the summer reliability period, the analysis that establishes the analytical limit for the main steam tunnel temperature corresponding to a 25 gpm leak has been refined to remove unnecessary conservatism and raise the analytical temperature limit.

The Analytical Limit (AL) is the point where desired action is to be initiated to maintain the integrity of physical barriers or other plant characteristics which must remain intact or operational. The Allowable Value is chosen based on an Analytical Limit that is calculated to detect a leak equivalent to 25 gpm.

A new Analytical Limit of 194.77 OF was determined based upon a refined analysis that established the analytical limit for the main steam tunnel temperature corresponding to the same 25 gpm leak, The previous Analytical Limit was 154 OF. The determination of the new analytical limit is discussed below in section 4.1.

The corresponding setpoint calculation determined a new limiting Allowable Value of 5183 OF based on the new Analytical Limit of 194.77 OF. The Allowable Value of <1 83 OF was derived from the Analytical Limit by correcting for calibration, process and other instrument uncertainties as discussed in Technical Specification Bases 3.3.6.1 and USAR Section 7. The new Nominal Trip Setpoint of 173 OF was determined considering instrument drift, with the drift value for the 24 month fuel cycle, and loop uncertainty.

The determinations of the setpoints are discussed in sections 4.3 and 4.4 below.

The Emergency Plan is affected by the change to the main steam tunnel temperature trip setpoint because Emergency Implementing Procedure EIP-2-001, "Classification of Emergencies" contains this setpoint in an Emergency Action Level and will require NRC approval. The discussion and justification is included in Attachment 4 of this submittal.

4.0 TECHNICAL ANALYSIS

4.1 Revised Analytical Limit As part of this amendment request, the analytical limit for the main steam tunnel temperature - high is increased from 154°F to 194.77°F. The basis for the current main steam tunnel temperature - high setpoint analytical limit is an area temperature rise equivalent to a 25 gpm steam leakage rate as is described in RBS USAR Section 5.2.5.1.3, "Detection of Leakage External to Containment". The basis is unchanged by this request.

RBG-47157 Page 3 of 11 It should be noted that the main steam tunnel temperature - high isolation function is not credited in any analysis reported in the USAR.

The following summarizes the principle differences between the current analytical limit calculation and the proposed analytical limit calculation:

Computer Program Used:

The current calculation was performed using the THREED computer program, a Stone &

Webster modified version of RELAP4 intended for sub-compartment analysis. The proposed calculation was performed using the Generation of Thermal-Hydraulic Information for Containments (GOTHIC) computer program. RBS has used GOTHIC for analysis of sub-compartments in containment as well as the analysis of high energy line breaks in the Auxiliary Building as described in submittals to the NRC dated May 14, 2002, supplemented by letters dated June 27 and July 9, 2003, April 7 and May 12, 2004, and approved in Amendment 139 to NPF-47 dated May 20, 2004."

Steam Tunnel Unit Cooler Simulation:

The current calculation simulates cooling in the steam tunnel by using two boundary conditions. One boundary condition represents the suction to the unit cooler providing cooling for the steam tunnel and the other represents the unit cooler return air. A fan is used to simulate the unit cooler flow, drawing air from the steam tunnel and to simulate the return air flow. This arrangement returns air to the leak volume at a constant temperature of 1050F and 32% relative humidity, regardless of the conditions of the air being drawn from the leak volume. This approach is highly conservative as it ignores the fact that as the suction temperature and humidity rise the return temperature will also rise. Thus the calculated temperature at one hour is artificially low, which is conservative for this application.

The proposed calculation explicitly simulates the unit cooler using the AIRCOOLER component in GOTHIC. The AIRCOOLER model in GOTHIC calculates both sensible and latent heat removal. Inputs for the AIRCOOLER were selected to represent design performance of the unit cooler.

This change in steam tunnel unit cooler modeling is the main contributor to the increase in the analytical limit.

Initial Conditions:

The current analytical limit is based on an initial temperature in the steam tunnel of 128°F. The proposed calculation uses a value of 11 0°F, which is based on - 2 years of plant data and represents a bounding low value during power operation.

The proposed calculation uses a value of 60°F for the service water temperature which is the cooling medium for the unit cooler. There is no similar input in the current calculation method.

RBG-47157 Page 4 of 11 Low values for initial area temperature or service water temperature will yield low values for the area temperature at one hour, which is conservative. Selecting bounding low values for the initial temperature and service water temperature will result in a conservative setpoint applicable to foreseeable temperatures.

Conclusion As identified above, this function is not credited in current design basis analysis. Also, the proposed changes maintain the 25 gpm design criteria. As a result, there are no changes to other design basis evaluations, plant response or consequences to an event.

4.2 Current Licensing Basis Review for the Main Steam Tunnel Ambient Temperature Isolations.

USAR Section 5.2.5 discusses the reactor coolant pressure boundary (RCPB) and ECCS leakage detection system.

Section 5.2.5.1, "Leakage Detection Methods," identifies the system is designed to be in conformance with Regulatory Guide 1.45 and Reference Section IEEE 279. This proposed change to TS does not change the conformance to IEEE 279. Regulatory Guide 1.45 deals with leakage inside containment and is not applicable to the proposed change.

USAR section 5.2.5.1.3, "Detection of Leakage External to Containment", describes that main steam tunnel ambient temperature is used as an indication of RCPB leakage and that the temperature setpoints are predicated on an area temperature rise equivalent to reactor coolant leakage into the monitored areas of 25 gpm. The proposed change does not change the basis for these instruments. The setpoints are still predicated on an area temperature rise equivalent to reactor coolant leakage into the monitored areas of 25 gpm.

USAR section 5.2.5.2, "Leak Detection Instrumentation and Monitoring", states that high temperature alarms in the main control room provide a signal to close the main steam and drain line isolation valves, RCIC steam isolation valves and the RWCU system isolation valves. This description remains unchanged. USAR Tables 5.2-7 and 5.2-8 indicate that high main steam tunnel temperature provides both alarm and isolation for the Main Steam, RCIC and RWCU systems. The proposed change does not change the number of temperature elements employed or where the instruments are physically located. The proposed change does not delete or remove any high temperature alarms.

USAR Section 5.2.5.1.3 "Detection of Leakage External to Containment" discusses the areas outside the containment which are monitored for primary coolant leakage equipment areas in the auxiliary building, the main steam tunnel, and the turbine building. These monitors have temperature setpoints which are predicated on an area temperature rise equivalent to reactor coolant leakage into the monitored areas of 25 gpm. The proposed change does not change the basis for these instruments. The setpoints are still predicated on an area temperature rise equivalent to reactor coolant leakage into the monitored areas of 25 gpm. The proposed change dictates a change in RBG-47157 Page 5 of 11 the Allowable Value for the main steam tunnel temperature isolation instrumentation due to a change in the engineering model being used. As stated above, the GOTHIC model has been previously used for River Bend Station calculations and is being applied for determining the Analytical Limit for these instruments. The Allowable Value is then determined from the Analytical Limit using the existing plant setpoint methodology.

USAR section 7.1.2.5, "NSSS and BOP Safety-Related Set Point Methodology", states that the methodology used for determination of NSSS safety related setpoints is documented in General Electric document NEDC-31336 that was approved by the NRC in 1993. The change to the setpoints is done in accordance with this methodology; therefore, no change to the USAR or other licensing basis is needed. Further discussion of the changes to the setpoint using the approved method are included below.

USAR section 7.6.1.2, "Leak Detection System" describes that the safety related portion of the leak detection system includes main steam, RCIC and RWCU system leak detection and that one of the signals used to initiate alarms and isolation is main steam tunnel area high ambient temperature. As previously stated the proposed change does not affect the design or operation of the main steam tunnel ambient temperature instrumentation. In addition, the setpoint will continue to be based on present licensing basis criteria.

USAR 7.3.1.1.2, "Containment and Reactor Vessel Isolation Control System (CRVICS),"

discusses the instrument channels, trip logics, and actuation circuits that automatically initiate valve closure providing isolation of the containment and/or reactor vessel, and initiation of systems to limit the release of radioactive materials. This USAR section also discusses variables which provide inputs to the CRVICS logics for initiation of reactor vessel and containment isolation, as well as the initiation or trip of other plant functions when predetermined limits are exceeded. As previously stated the proposed change does not affect the design or operation of the main steam tunnel ambient temperature instrumentation. In addition, the setpoint will continue to be based on present licensing basis criteria.

4.3 Allowable Value Determination - River Bend Station (RBS) Setpoint Methodology The RBS setpoint control program is implemented utilizing engineering process controls, plant procedural controls, and the corrective action process. This program ensures the associated instrument channel is capable of performing its specified safety functions.

USAR Section 7.1.2.5, "NSSS and BOP Safety-Related Set Point Methodology," states that the methodology used in determining BOP safety system set points is in accordance with Regulatory Guide 1.105, Revision 1 and IEEE 279-1971. It references USAR Table 1.8-1 for the River Bend Station position on Regulatory Guide 1.105. It also states that the methodology used in determining NSSS safety system setpoints is similar to that described above for BOP safety system setpoints. This methodology is documented in NEDC-31336, General Electric Instrument Setpoint Methodology, dated October 1986.

This document was developed by the Instrument Setpoint Methodology Owners Group (ISMG) and approved by the staff on February 9, 1993. The proposed revision does not change any requirement for setpoint methodology.

RBG-47157 Page 6 of 11 This methodology was recently discussed with the NRC staff during the staff review of the RBS 24 month cycle submittal and approval. In a letter dated August 17, 2010 (ML102350155) Entergy responded to an NRC question (RAI) concerning compliance with RIS 2006-17. As stated in this response the original Nominal Trip Setpoint, Allowable Value, and Analytical Limit design bases were established from the supplier design requirements. Calculations were developed approximately 10 to 15 years ago to confirm that the plant conditions (measurement and test equipment, device accuracies, drift, etc,) were conservative relative to the assumptions in the design basis.

These original calculations, subsequent revisions and the methods of calibration continue to confirm that the existing Technical Specification Allowable Values and Technical Requirement Manual Nominal Trip Setpoints are conservative with respect to the original design basis. Figures. and Tables supplied with the response summarized the important values created by or relative to sample setpoint calculations for the 24 month submittal.

The conclusion to the RAI response was that the tables and charts demonstrated that the setpoint calculations and associated surveillance procedures affected by the change to a 24 month fuel cycle at River Bend Station met the intent of NRC Regulatory Issue Summary 2006-17. Amendment 168 to the River Bend operating license for the 24 month cycle was issued August 31, 2010 (ML102350266).

The methodology described in the August 17, 2010 letter has not changed as a result of this request. The only change is the computer code being used to determine the analytical limit and, as a result, the corresponding setpoint values using the approved methodology. Therefore the proposed change is not a change to the setpoint methodology.

4.4 Nominal Trip Setpoint The RBS setpoint control program also determines the nominal trip setpoint. This program contains the same engineering process controls, plant procedural controls, and the corrective action process used for determining and controlling the allowable value.

The proceeding discussion of license basis and setpoint method described above also applies to the nominal trip setpoint. In addition the as-found and as-left values are identified in the setpoint calibration procedures. Setpoints found outside of the established as-found limits are identified in the corrective action process.

The nominal trip setpoints are identified in the Technical Requirements Manual (TRM).

Changes to the Nominal Trip Setpoint are controlled under 10 CFR 50.59.

RBG-47157 Page 7 of 11 4.5 Equipment Qualification Equipment qualifications have been reviewed within the affect area and determined the equipment will perform as required with the revised setpoints.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria Section 50.36 to Title 10 to the Code of Federal Regulations (10 CFR) Part 50, requires that TS include limiting conditions for operation (LCOs) for any structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. Section 50.36 to 10 CFR, also requires that TS Surveillance Requirements (SRs) be requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCO will be met. When an LCO for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the condition can be met.

Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, Criterion 13 Instrumentation and Control states that:

Instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems. Appropriate controls shall be provided to maintain these variables and systems within prescribed operating ranges.

USAR section 3.1.2.13 discusses the RBS conformance to this General Design Criteria.

The proposed changes do not affect the conformance as presented in this USAR section.

Appendix A to 10 CFR 50, Criterion 24-Separation of Protection and Control Systems states that:

The protection system shall be separated from control systems to the extent that failure of any single control system component or channel, or failure or removal from service of any single protection system component or channel which is common to the control and protection systems leaves intact a system satisfying all reliability, redundancy, and independence requirements of the protection system. Interconnection of the protection and control systems shall be limited so as to assure that safety is not significantly impaired.

USAR Section 3.1.2.13 discusses the RBS conformance to this General Design Criteria.

The proposed changes do not affect the conformance as presented in this USAR section.

RBG-47157 Page 8 of 11 Regulatory Guide (RG) 1.105, Revision 1, Setpoints for Safety-Related Instrumentations, describes a method acceptable to the NRC staff for complying with the NRC's regulations for ensuring that instrumentation setpoints are initially within and remain within the TS limits. USAR Table 1.8-1 and Section 7.1.2 discuss the RBS conformance to this Regulatory Guide. The proposed changes do not affect the conformance as presented in this USAR section.

Regulatory Issue Summary (RIS) 2006-17, "NRC Regulatory Issue Summary 2006-17 NRC Staff Position On The Requirements Of 10 CFR 50.36, "Technical Specifications,"

Regarding Limiting Safety System Settings During Periodic Testing And Calibration Of Instrument Channels,",dated August 24, 2006, discusses the requirements of 10 CFR 50.36 related to Limiting Safety System Settings (LSSS) and provides an approach acceptable to the NRC to address LSSS issues. The LSSSs are settings for automatic protective devices related to those variables having significant safety functions.

As discussed above in Section 4.3, compliance with this RIS has previously been discussed in response to a Request for Additional Information (RAI) during the review of the RBS 24 month cycle submittals. This resulted in part in the approval of Amendment 168 to the RBS facility operating license on August 31, 2010 (ML102350266).

5.2 No Significant Hazards Consideration Entergy has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," discussed below.

The requested change would affect certain Technical Specification (TS) Allowable Values for Main Steam Tunnel Ambient Temperature isolation instrumentation and the Emergency Action Levels supporting the Emergency Plan.

1.

Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change increases the Technical Specification allowable value for the main steam tunnel ambient temperature isolation instrumentation for the main steam line isolation, Reactor Core Isolation Cooling System isolation and the Reactor Water Cleanup System isolation. This TS change does not introduce the possibility of an increase in the probability or consequences of an accident because the basis for the instrument setpoint is not being changed as a result of this request. The proposed TS change involves no physical alteration of the plant. The proposed TS change does not degrade the performance of, or increase the challenges to, any safety systems assumed to function in the accident analysis. Therefore, the proposed change does not involve a significant increase in the probability of an accident previously evaluated.

RBG-47157 Page 9 of 11 The consequences of a previously evaluated accident are not significantly increased. The proposed change does not affect the performance of any equipment credited to mitigate the radiological consequences of an accident.

The basis for the main steam tunnel ambient temperature isolation instrumentation has not changed as a result of this proposed Allowable value change.

The proposed change to the Emergency Action Level (EAL) does not increase the probability of an accident. The change only impacts the initial condition for entry into the Emergency Plan and thus has no impact on the probability of an event. The proposed change to the Emergency Action Level (EAL) does not increase the consequences of an accident. As described in the Technical Analysis the revised setpoint continues to support the current licensing basis and event analysis.

Because the process, personnel, and equipment involved in implementing the Emergency Plan would complete the same functions as those completed under the existing Emergency Plan, the plan would continue to ensure adequate protection of public health and safety.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

As discussed above, the proposed change involves increasing the TS allowable value for the for the main steam tunnel ambient temperature isolation instrumentation for the main steam line isolation, Reactor Core Isolation Cooling System isolation and the Reactor Water Cleanup System isolation. The proposed TS change does not introduce any failure mechanisms of a different type than those previously evaluated, since there are no physical changes being made to the facility. No new or different equipment is being installed. No installed equipment is being operated in a different manner. The computer program being used has been previously used and reviewed. As a result, no new failure modes are being introduced. There are no new types of failures or new or different kinds of accidents or transients that could be created by these changes.

The change affects the implementation of the Emergency Plan by changing the EALs temperature value for entry into the Emergency Plan; however, the basis for the temperature value is not changed. The change to the EAL does not impact any plant equipment or systems needed to respond to an accident, nor does it change the results of an analysis of plant accident consequences.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

RBG-47157 Page 10 of 11

3.

Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

As discussed above, the proposed change involves increasing the TS allowable value for the for the main steam tunnel ambient temperature isolation instrumentation, the main steam line isolation, the Reactor Core Isolation Cooling System isolation and the Reactor Water Cleanup System isolation. The effect of this change on system availability is not significant, based on the determination that the basis for the allowable values is not being revised. The proposed change does not adversely affect the condition or performance of structures, systems, and components relied upon for accident mitigation. The proposed change does not result in any hardware changes. Existing operating margin between plant conditions and actual plant setpoints is not significantly reduced due to these changes. The proposed change does not.significantly impact any safety analysis assumptions or results.

The change to the Emergency Plan does not reduce the margin of safety currently provided by the plan. As discussed in this submittal the change does not revise the design criteria of detecting a 25 gpm leak. Also the methods used to determine the revised analytical limit and setpoint values are currently' accepted. The proposed change does not impact other design basis evaluations or consequences. Therefore the changes do not affect a margin of safety identified in the plant accident analysis.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

RBG-47157 Page 11 of 11 6.0 Precedence Similar request to revise instrument setpoints has been identified as follows:

1 Pilgrim Nuclear Power Station," Request for Technical Specification Change Concerning Change of Trip Level Settings, Calibration Frequencies, and Editorial Changes, Revision 1," Dated October 10, 2002 7.0 References 1

River Bend Station, "License Amendment Request (LAR) 2001-43, "High Energy Line Break Analysis Method," Dated May 14, 2002 2

General Electric Instrument Setpoint Methodology, NEDC-31336, October 1986.

3 River Bend Station, "Response to Request for Additional Information on License Amendment Request 2009-05, 24-month Fuel Cycles," Dated August 17, 2010 4

NRC, "River Bend Station, Unit 1 - Issuance of Amendment RE: Revise Technical Specification Surveillance Requirement Frequencies from 18-to 24-Month Fuel Cycle Interval (TAC No. ME1 872)," Dated August 31, 2010 RBG-47157 Technical Specification Markup RBG-47157 Page 1 of 3 Prinary Countainment and Drpvwe Isolation Instrumentation 3.3.6.1 T73b 3.3.6.1-1 (1I f5)

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SR 3-1r.1.1 SR 331..12 SR 3111.5 SR 311I6.5 F

SR 3-1..1.1 SR 31,1.2 SR 1.1.6,1.3 SR 3-1.,15 SR 3-1r.16.5 F

SR 3-1r.1.1 SR 3116,12 SR 3.1161.3 3R 311,.1q SR 1*&16S 0

SR 11615J 5:2955 t: 16SAF 5: 163.LSF NA 5: 121.I-F 5:64-2 Indmvg~

& 1.689 ni NA

  • RJV BEND 3.3-55 Amendment No. 94 1

RBG-47157 Page 3 of 3 Primary Containment and Drjwel Isolatio Instrumentation 3.3.6.1 Tabt 3.236.1-1 (psJ atS)

MFlsry" C~ofMmfl ad Dwtell Iallma Insbfm l

A=PLICAELE OONOITlcJS MOOEDi OR REF*R*CED OTHER ROQUUIU FROM UFEC=IED CHANIN.S PER REUAIREO SURVELSANIC A.LOWABLE FU NCTICIN CONDIM TRP SY*'EM ACTON r1 RIEOU[RfhOITS VALUE A.Rs R,: at" Cleaup ~aa arnm tolala

a.

Ol**amFI~-Hlgf 1.2,.3 1

F 23.3.6.13

,62.1 en SF R 3.3...12 ER 3.3.r.12 ER 3-3.6.15 IL. OleRflSFli-Tnp-l 12.3 1

F SR 3-3.E.12

  • C47 sermbf ER 3.3..1 SR E

.R. 1.,6 C

RWCU Heut Erlmw 12.3 I

F SR 3J.1I-I07.'F EBArnle Rnom SR 3.3.1,2 Tevera-le-fghn SR 3.6.1-S SR 3.3..1.

d.

RWCUIPumprvana 1,23 I

F ER 3.3.6.1.1 L 16.F Teer~xald-Nlgh ESR 3.3.6.12 ER 3.3.6.1 SR 3.3.6.1,5

a.

RWGCUYeN*MRoms 12,3 I

F ER 3-3.6.1-1 I11."rF Terve'she-High ER 3.3.5.12 SR 3.3.6.15L ER 3.3.1-1-6, t

RW CUDemilTal' Rom 123.

F SR 3.+/-.13

  • u Ir.SF TVarVemalwr44gtl ER 3.3.6.12 ER 33.6.15 SR 33.6.15 RWCU Reed Tt Room 12.3 1

F ER 33..1.

!*11*.tF Te,*v*

as H

E*R 33.6.11 SR 3.3.6.1J ER 33.6.1.

.L R in aEsr LIM '**,*

123 I

F ER 3-3.15.1-1 18343F AL"-fliTempuflnu.m-i ER 3-3.6.12 ER 3.3..1.5 ER 3-3.6.15L L.

Rae-U fls'atu Wa123 2

F ER 3.3.10.1 Z - 47 bfln LenH-Aw UM. Loeve 2 ER 3.3.&.12 SR 3.3.6.1.3 ER 33.1.1-5 SR 3-3.6.156 J.

E*tfl LlqzM CaitS Ss.n 1,2 1

SR 3.3.6.1-6 NA IL k*tUml hllan 1,2.3 2

G SR 33.6.1, NA RIVER BEND 3.3-56 Amendment No. 14 RBG-47157 Technical Specification BASES Mark-up (For Information Only)

RBG-47157 Page 1 of 1 Primary Containment and DryweI Isolation Instrumentation B3.3.6.1 BASES SURVEILLANCE SR 3.3.6.1.3 (continued)

REGUIREMIENTS Table 3.3.6.1-1. If the tip setting is discovered to be less conservative thmn accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety analysis. Under these conditions, the setpoint must be readusted to be equal to or more conservative than accounted for in the appropriate setpoint methodology.

The Frequency of 92 days is based on the reliability analysis of References 5 and 6.

SR 3.3.6.1.4 and SR 3.3.6.1.5 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. For Functions I.e.,

3.f and 4-h there is a plant specific prooram which verifies that the instrument channel functons as recuired by verifno the as4eft and as-found setting are consistent with those established by the setpoint methodology. CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifs between successive calibratons consistent with the plant specific setpoint methodology.

The Frequency of SR 3.3.6.1.4 and SR 3.3.6.1.5 is based on the assumption of the magnitude of equipment drift in the setpoint analysis.

SR 3.3.6.1.6 The LOGIC SYSTEM FUNCTIONAL TEST demonstrates the OPERABILITY of the required isolation logic for a specific channel-The system functional testing performed on PCIVs in LCO 3-6.-13 and on drywell isolation valves in LCO 3.6.5.3 overlaps this Surveillance to provide complete testing of the assumed safety function. T-he 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power(

(continued)

RIVER BEND B 3.3-167 Revision No. 14._. I RBG-47157 Emergency Planning Evaluation RBG-47157 Page 1 of 9 Emergency Planning Evaluation The "Main Steam Line Tunnel Ambient Temperature - High" trip setpoint value in the Technical Requirements Manual (TRM) Nominal Trip Setpoint (NTSP) specification is being changed from 1440F to 1730F. When temperature sensors reach this setpoint a signal is sent to actuate the Control Room annunciator "Main Steam Line Tunnel Ambient Temperature - High" and a signal is sent to close the Main Steam Isolation Valves (MSIVs), Reactor Core Isolation Cooling (RCIC) steam isolation valves, and the Reactor Water Cleanup (RWCU) system isolation valves on the piping penetrations running outside the containment barrier into the Main Steam Tunnel.

The "Main Steam Line Tunnel Ambient Temperature - High" setpoint is based on an area temperature rise equivalent to steam leakage into the steam tunnel of 25 gpm. As discussed in Attachment 1, an updated calculation determined a revised Analytical Limit (AL) of 194.77 'F for the 25 gpm steam tunnel leakage by using more realistic assumptions and removing excessive conservatisms. Also, as discussed in Attachment 1, an updated TRM NTSP calculation determined a limit value of 173 'F for the "Main Steam Line Tunnel Ambient Temperature - High" trip function and a Technical

.,.Specifications Allowable Value (AV) limit of <183 *F.

This plant change affects the Main Steam Tunnel NTSP temperature value (1440F to 173 0F) listed for Fission Product Barrier IC-EALs, Reactor Coolant System (RC) Barrier RC3 (RCS Leak Rate) "Loss" and "Potential Loss" determination. The RC3 Bases uses the Main Steam Tunnel isolation setpoint as the bases for determining a reactor coolant leak outside containment in determining the EAL entry condition.

RC3 Bases (Excerpts, See Enclosure A for entire bases)

LOSS - An unisolable Main Steam Line break represents a loss of the reactor coolant system barrier. This EAL is included for consistency with theALERT emergency classification.

The leak is NOT isolable from the Main Control Room OR an attempt for isolation from the Main Control Room panels has been made and was not successful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification, this INITIATING CONDITION is not applicable. Dispatch of operators outside the Control Room for manual attempts to close the valve is not considered.

POTENTIAL LOSS - The alarms for the high area ambient temperature are associated with the TS 3.3.6.1 allowable values for primary containment isolation. The T.S. allowable high ambient temperature setpoints are set low enough to detect a leak equivalent to 25 gpm. The use of an isolation alarm and EOP condition provides a method of rapid identification of the EAL condition without the task of obtaining values from a back panel indication. [Note: the TS 3.3.6.1 new temperature value is 183 0F and the TRM NTSP value with design margin is being changed from 1440F to 173 0F.]

RC3 Loss bases specify that an attempt for isolation from the Main Control Room should be made prior to the accident classification. This action would be performed after receiving the high temperature isolation signal. The temperature isolation trip setpoint RBG-47157 Page 2 of 9 will now occur at 1730F. RC3 Potential Loss bases specify the setpoint is set low enough to detect a 25 gpm leak and use of the isolation alarm (will be set at 173 0F) provides a rapid identification of the EAL entry condition. The new higher setpoint still provides conservative indication of the 25 gpm leak. The new higher setpoint could slightly delay the declaration of the event due to the temperature change from 144 to 1730F. However, it does not delay declaration beyond the 25 gpm limit that is in the RC3 bases. The temperature in the EAL IC was chosen because they match the isolation and alarm setpoints which provide positive indication of the entry condition.

This change preserves that condition. The RC3 Main Steam Tunnel EAL determination temperature value should be the same value as the isolation and alarm setpoint value.

Changing the RC3 Loss arid Potential Loss Main Steam Tunnel temperature from 1440F to 173 0F does NOT decrease the effectiveness of the EALs since the new temperature value will be the isolation and alarm setpoints for a 25 gpm leak in the main steam tunnel as described in the EAL bases. The RC3 EAL Main Steam Tunnel numerical temperature value should be equal to the Main Steam Tunnel isolation setpoint value to ensure the value will detect a 25 gpm leak, EAL temperature matches isolation setpoint and Operators will receive an isolation alarm to readily identify potential EAL entry conditions.

The increase in the "Main Steam Line Tunnel Ambient Temperature-High" trip function temperature from 144 OF to 173 OF will help minimize the possibility of a spurious instrument trip to isolate yet maintain the ability to detect and isolate based on a leak of 25 gpm in the Main Steam Line Tunnel.

This increase in the setpoint for "Main Steam Line Tunnel Ambient Temperature-High" trip function and alarm will still maintain the plant within current design basis for the detection of 25 gpm leaks in the Main Steam Line Tunnel.

The temperature value change does not impact the requirement for recognition and classification of an emergency within approximately 15 minutes of the time indications are available. The temperature change still maintains the capability to identify conditions and does not degrade the availability of the approved EAL threshold indications. The temperature value change does not affect the time requirement or capability to perform in a timely manner the functions for notification, protective action recommendation, and alert and notification of the public.

RBG-47157 Page 3 of 9 REFER'CE USE AflACnMENTy 3 PAGEcOF z

.-*+/-qgZ ZZ FISSION PRODUCT BLARIER

_m i

.fF RGF %I¶EUIA I

U' I

FGL LJLhI i

Loss ofazy two baemss and loss at paotnnal loss af *trd Loss o pololil Loss of any two ban'm

-n me EmnMrancvActiom Loestk (1)

E.

l mU..MCBAKiORLU.dt',l, (n)

L L..

Lsor pom-a lo~ss of my nio b-d~s Il.

L.oss of an tw*,y I

.1

-AND

.3 Lss orpol onlo loss oaf tid b*rrior LIJ1l)lZQlBI I

FAI LIII-i-II FM' llldldlh Amy Ios or anzy polmsiml loss of eitbrr fsid clad or RCS E

er.-

Arbctonu Le. fsk: (1)

I Ay loss or any patedal loss of flhel.ld a&

Amy loss or sany poumial Loss of RCS Any loss or my potnni loss ofcu Emerammt Action Led-*

): (1)

I.

Any Loss or any pottal loss of rmaimst rumL CILAn (FC) 0o--

REACTORL COOLANT SY~tEM (KC)

RB njJPýlIK CON.TAIN"-iLE (PC) B Lnn PureE Loss.Pmo n

Los LoL L-.

L-s.

P L.-is Fat F-wzy coolan Coolant.'n-y fle CI Eko.U Iou.f~l n~

as-Na.

PCI lbsooaaapsn.R~pdnov no Los..

o-f PC pan L15 puadiSsn 1

g PC hy,-op00a F

son týJ--jqs I

a.,HO-]OL cm'v.

ra Bsn.sumi R~Jp L-Loss tiu RFl' 3-1 bonE loss th I

L--S RsorFsV AMP nra los o.ta 12 MoFSRa

-... 1 nr

-1ou Na.

mPC Snodcog onr Loc--WRb 12ba f

os!tmswtittafpctu SAP-! ad LAP-2

,t oln t

Cooint=y C

rl

.,-m XCI IC LO-h IwoUnlC 1

nbn.

!SIoa aasa Mi FC 7-27cýthso F.b-c f With Cm-on rotanon soon.-

inoot NO-RE1d ro.g wr S

d3ý0OO 3LA.h bun& - ntdkasd by di.

AND (P60l-19IL-A2)

=

-I-taw IaS-nT-s o ".e I75'n o..x unIabl. F....y.3ya

-- k Wds PC -

nomby n.y a

l. nF2 dni san-b-

nialmo minuia ftut b31i

.. d on-wi toatona LlbnbnvaniSg pa lJOP.

=SAP.

a&

!Jnisolabln ItSIta

-0.if PC n-,ndb.ind by -~y aS =onpaws -

nu I

r_-

Idwd 1-1__

i.___

Tbl Fl-----------

-I e

tI XJC4 -- Y-1115-1520-cdiag par tb-100 Mg., d-. to -taco coolan bIrn.

4t ~tJO OO5 tNInc~

aiauoaC0=

H 18 10.000 Fb Wjn]!Ddg..at I Anymaisnat j

n atobmt S

p-fn.cRnm o

a Ay c dronj o pa PC5 WjdgmatIAycodamto-n

-yatnn lb. qpnt I pneo ErnryIopn f

i Bp yI fti ogapiwat oI p-yB-ais tn

-Enoyfon, oIos.EtiuayCu.yta oft tE la bns I

otanllssot~it.

XSI 3~

p In -m

-t.

A-R

~

ont bw iy..

t bi Dicura to tdmnsoIn.

Domno trints.f ndr.

mo oft.D

]UM iniatu oeh osot aiaIs f.puy Izian nn asot

_________I_________ý claiW fba.

Fiatt tAet.(uvrto h~ant Soa q.~xzhlomtoi I

Po,Opetha 2

Snu I

Itu i.*Lbdoý 4

CA-nnds 5

R5S*

0D lDcri't FI]P-2-001 REV - 020 PAGE 24 OF 169 RBG-47157 Page 4 of 9 REFlENCE USE ATTACEMENT 3 PAGEaOF2L2j 4

rk FISSION PRODUCT LARRIER T.ABLE FIL PCU Loso of Primy Cocutrnnment Pkrametw

-Aea TremlR'

...*l R-MambolnIa v

R.ER A arm 200rF 1213 9.5E-+93 M.Er RHB s wea 200F 1214 9.5E+03 aR..'r m*mC emt,ea "N

1215 9.5E+03 nmbr R

XIC J-!0r F 1219 9,E-Q3 n.hr

./2.

20(r F NA b*,LW.v PUM MGM I 20(r Fl"2q' (A) 2 (B)

TABILE F1 RC 3 Po*tentia Loss ofVRCS PainmA~~emerfw Axis Rzdiatiio L7ar ERR A equipoent ama 11i"p 1213 8.2E-401 mR'7i P60l-20A-B4) maR B equipmeixt ano 117'F U214 9.2E+01 mR'Jr C9601-20A-B4)

E1 MeC eq amsta N/,A 1215 9.2E-01 mbR~

ECC ar trp 1219 1I20E+02lmRl (P601-PA-A/3B1:B)

(P53-IA-Aif2)

-L o a

.. b&'

EIP-2-GO1 REV - 020 PAGE 25 OF 169 RBG-47157 Page 5 of 9 iFl"3UENE LSM FAL BASES RC3 REACTOR COOLANT SYSTEM flgey-clim&.Le;vl RCSlmekaip LOSS:

UT.us.lae mtn am lin break as *indamd by the fun of both &MSVs in any cu lb. tomose High MSL Low aimimd (P6O1-19A-A2) 3.:n Steam ibm-iT-Mu gpsi than iZA". --------------

POTENTIL LOS& RCS ]eakp g-ar tbam 50 gpmb 1-6 d* dtyweil Unso.ble pmsy sytelesk outdie PC as inatedy ay*

armes tmnp-ame alum a

r am rranu level.,I-m Table F2 KC 3 Pa~eaualls SiNGS PrtMMs Aia

]Tepnfi 1 Asa2R diL.

"1t Cpealis Valna KRfAairh ai1ry 1213

&2 11+01 mIr/

(P601t/20AI804)

RMLhBhqii arma IIryF 1214

&2 E-01 mIL'kr Xf.l¢C aqunt -'

24.

1233 1_211-41 m*"l" RC rm=

1f1F 1219 1.20.E12 md~r (N0-19VAI, 1i AOn. B0, RX_

iPp

-O2J I

AA)2 nd2 W1 36Y F

.d6rdPtWAA0 aad

_J-a o

,,144 EIP-2-401 "REV- 020 PAGE 76 OF i69 RBG-47157 Page 6 of 9 REFERENCE USE ATTACIM =ET 8 PAGE 42 OF 118 EAL BASES LOSS - An unisolable Main Steam Line break represents a loss of the reactor coolant system barrier.

This EAL is included for consistency with the ALERT emergency classification.

The leak is NOT isolable from the Main Control Room OR an attempt for isolation from the Main Control Room panels has been made and was not successful. An attempt for isolation should be made prior to the accident classification. If isolable upon identification, this InTIATING CONDITION is not applicable-Dispatch of operators outside the Control Room for manual attempts to close the valve is not considered POTENTIAL LOSS - A reactor coolant system leak rate ofgreater than 50 gallons per minute is at a level indicative of a small breach of the RCS but which is well within makeup capability of normal and emergency high pressure systems. Core uncovery is not a signifcant concern for a 50 gm leak; however, break propagation leading to a significantly larger loss of inventory is possible. A leak of this size is a precursor of the loss of the reactor coolant system integrity and is therefore considered to be a potential loss of this barier.

If the leak detection system leak rate information is unavailable (ite-, LOCA isolation, loss of power),

other indicators of RCS leakage should be used. Other indications include a rise m drywell temperature and pressure and a rise in the drywell radiation monitors-If the leakage computer is unavailable, sump level and pump status may help determine if greater than 50 gpnt If the DFR discharge line containment isolation valves have nzt isolated and a pump is running continuously without lowering sump level, the leakage may be assumed to exceed 50 gpm The second pump can be started to verify that the first pump is not degraded-It is not intended to conclude a potential loss of the RCS baffer if both pumps are degraded and the observed leak rate as noted by rate of rise of level in the sump or calculated by the computer is such that it clearly confirms leakage below 50 gpm_

A VALID indication of area temperature(s) greater than or equal to the system MOV Tech Spec isolation value or area radiation level(s) greater than or equal to the monitor high alarm resulting from a primary system discharging into the Auxiliary Building is indicative of conditions inwhich significant RCS inventory is being lost This is therefore considered to be a potential loss of the reactor coolant system boundary The area radiation values are consistent with the EOP maximum normal operating values. The alarms for the high area ambient temperature are associated with the TS 3-3-6-1 allowable values for primary containment isolation. The T.S allowable high ambient temperature setpoints are set low enough to detect a leak. equivalent to 25 gpnt The use of an isolation alarm and EOP condition provides a method of rapid identification of the EAL condition without the task of obtaining values from a back panel indication.

References:

EIP-2-fO1 RFX" - 020 PAGE 77 OF 169 RBG-47157 Page 7 of 9 REFERENCE USE AT"AC-IMBfT 3 PAGE 1 OF 2 FISSION PRODUCT BARRIER 1FG1 1173111 irs*

imniia n

Al LFIJs U~I'm-I Loss of any two barntams and loss or potential loss of thrd E.emnenc' Aotm LeveACs: (-)

I.

Loss oKf-ay two bsrners As*

L~oss. or potential loss of third barrier L.os or potential loss of any tawo banims

1.

Loss or potential loss of any two baner Any lsm or ay potential toss of either fitel clad cKrRCS EnSEresAr

&fo X=vhs (I)

1.

Any boss or any potenti lass of fDa clad OR Any loss or any pat-rWial loss of RCS Any Doss or any potential loss of conainnt 1..

Any* soan-Wotetl loss (1)

I-Amy loss or any potential loss of condtnn FUEL CLAD Q

I.-R E[ACWR COa0LAir sinns (MC) Ba*er,

-flfMARY CONAnI NMN 0" B.)

Pnane Loss Pot enial Less Patonn Loss potenial Loss Panalber Loss Potonial Los FC1 Primazy coolant Coolan activiy None RCI Drywell presst Drywall paesne greate ihm None PCI PaintarycoyentMpressnt Rapd tme dlossiof PC essure 15 psig and rimg trinity level greater tdta 300 pCegea 11S6 psin with indic-ations of preesnue following initial OR doseeqcual,*l -13s l urtr coolant eak inMdaYweL wessuze n*-

PC hydtga, in thla it e

a awe of NIE o.em PC plesurae response tot 2L can t with.LOCA con*itals DW hyshae cancemun FC2 Reacr vestse R3PV.r wea= lel less than RP.V wate level less xt1h RC, Reactor veswl RPV waer level less tbn -162 None PCI React'r vessel wat*r l]el NOn Entry ino PC looding waef level

-196 imbe

-162 inche wem level inches wit ion of Iwocetheres SAP-I and SAP-2 reactor coodan leak in drvwel FC3 Prinay Conmainmen radisme None RC3 RSLeAk Rate L-nisolzble main stem l RCSR eakage greatr than 50 PC3 Prainarycananmus Fan'r ofboat*vaher mmy None COMinlit menamn ERMS-REI6 break as idicated by fth epm inside lt fywell isolate faibrie or bypssed we line to dose and Mathanas readin greats thtan 3.000

&ime of bothMSVs in amy QA dnseam padswev to the n.,s L-r one line to close Unisolable pamtary system leak

,-vtinmentalit

_s AMaide PC as indLated by any aR ana tempetat t b-or arm Intutioal. nming per EOPs HPIIPAAZ r

l adiolion level alarm in or SAkPs

('P01-blA-A2)

Ta 3

pQ a

AM Unmslable R1CS leakage Main Stem ToMel aside PC as indiatedby any Teipsatuep g*rne area uMpwatxne of aim 113-F rad*iaton oe

.l s

ln TableFi1 RC4 Drywal radiatua Drywal radiation Mmtani Nonte PC4 Sigettana radioactive None Cootattinitti radtateanttowr RNMS-RE20 readfn greetr inltoy in tRtE 1MS-6E readiag hdan 1010 br ine to iftercactor

,cnamer gtear thnm 10.000 Rb.r coolamn leakge FC4 EDyldtent Any condi-ton in the Anycretnitian RCi ED jidgmein Any condio in the opinion Any ondmin the opinon of PC5 EDjudgnn Any conditno Jn the opnion A

condion i die op*um*

opinion of rhe Emergncy opiion ofthe Entrgency of the Emueruy Director i the Emargency DieCMFto tf the Emergency Dimect that of Directrtat Director that inbc*ateso DiLrectr t*i*ica-te I

loss of he

  • CS indicates potential loss off tUdIcates loss Of 1e Mry in.awtespotWialo ssofthe ofe she elcla*d am:i potential loss of she fitl barzier

-CEbLS cambinarrni earer ihi aa, n

-b-im dad baMer art (shim be n'

aplaba I

Pew Gyst-e 2

Sntw 3

HtMM&de 4

ColA tden 5

R-f-t Dt.&.

EIP-2-CO1 REV- 020 PAGE 24 OF 169 RBG-47157 Page 8 of 9 REFERENCE USE AflACI*--T 3 PAGE 2 OF 2 FISSION PRODUCT ELARRIER TABLE Fl PC 3 Loss of Primary Containment Parimeter Area Temuperstu"r e

Eacite

][wd

[snA euEment alnu 20 F

1213 9.5E03 mtb*r REEK B eganer rea 20WF 1214 9.SE-t03 mtb,'r RH. C equipment a rea A

1215 9.5E+03 mlPLr RCWC room 20(r F 1219 9.5E+03 mhbr

.MLSL Tunnel 2

F I/A

,Wa-pump room 1 200 F MA (A)2/

(B)

TABLE F2 RC 3 Potendil Loss of RCS P.Raem AreATes Rad]stioea Level M1rM va

  • rlme

.Gi RHRA equipmeunt ian117'F 1213 82.E+01 m.R/b (P601-20A-B-34 W.HMB equ4ipmi area 117rF 1214 8.2a-01 mR/hr (P601-20A-B4) sim C equpmEM area NIA 1215 8.2a-7R+01 DR RflC roam 1sr'F 1219 1_205+02 ma/hr (PO1-21A-2B6 M-SLTUmnM 173"F MA a*01-19A-AIA3.,]UB3)

RWCU

-pmp MeM I w A2 (B) 167 p NA I_

(PBO-La-A28 2)

I riP-i-00l REVr - 01' p

AGE 14, OF 160 RBG-47157 Page 9 of 9 REFERENCE USE A1TACI-r.hfT 8 PAGE 41 OF 118 EAL BAkSES RC3 REACTOR COOLANT SYSTEM Emergenc."Action Level:

RCS leak rate EAL Threshold:

LOSS:

Unisolable main steam line break as indicated by the falure of both MSIVs in any one line to close AND High MSL flow annunaator (P601-19A-A2)

AND Main Steam Tunnel Temperature greater than 173-F POTENTLAL LOSS: RCS leakage greater than 50 gpm inside the dchwell OR Unisolable primary system leak outside PC as indicated by any area temperature alarm or area radiation level alarm m Table F2 TABLE F2 RC 3 Potentia Loss f 0R-S P*uuemTae Area Radiation Level RHRA eqmu= area Il"F 1213 8.2 E+01 mRL.br (P601r0OA3,04)

RHR B equyipeff aea 117PF 1214 8.2 E+01 m]',Ir RER C eq*ipmx ara N/A 1215 8-E+01 mR.,br R._IC roam 8arF 1219 1.20 E-402 n3R'r (PGO1J21A'B~d)

MSL Tmnal 17Pr F N/A (POIAi19_AA01, A03, 1O1, B03)

RWCU pump room 1 (A)? 2 (B) 165I F N/A (P60/1A/A02 and W02)

Bases:

EIP-'-O01 REV1 - 020 PAGE 76 OF 169 RBG-47157 List of Regulatory Commitments RBG-47157 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document.

Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check one)

SCHEDULED ONE-TIME CONTINUING COMPLETION COMMITMENT ACTION COMPLIANCE DATE In addition to the identified changes to the X

Implementation Technical Specifications above, the BASES will be revised upon implementation to include the following information based upon TSTF-493 BASES revision to NUREG-1434, SR 3.3.6.1.5 to include the following statement; There is a plant specific program which verifies that the instrument channel functions as required by verifying the as-left and as-found setting are consistent with those established by the setpoint methodology.

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